View source for 05000341/LER-1996-017, :on 961025,discovered Multiple SRVs as-found Settings Outside of 1% Tolerance.Caused by Oxide Bonding of Pilot Valve Disc & Seat.Replaced Pilots W/Refurbished & Recertified Pilots W/Platinum Alloyed Valve Discs

{{Adams
| number = ML20199B152
| issue date = 11/12/1997
| title = :on 961025,discovered Multiple SRVs as-found Settings Outside of 1% Tolerance.Caused by Oxide Bonding of Pilot Valve Disc & Seat.Replaced Pilots W/Refurbished & Recertified Pilots W/Platinum Alloyed Valve Discs
| author name = Peterson N
| author affiliation = DETROIT EDISON CO.
| addressee name = 
| addressee affiliation = 
| docket = 05000341
| license number = 
| contact person = 
| document report number = LER-96-017, LER-96-17, NUDOCS 9711180231
| package number = ML20199B134
| document type = LICENSEE EVENT REPORT (SEE ALSO AO RO), TEXT-SAFETY REPORT
| page count = 9
}}
{{LER
| Title = :on 961025,discovered Multiple SRVs as-found Settings Outside of 1% Tolerance.Caused by Oxide Bonding of Pilot Valve Disc & Seat.Replaced Pilots W/Refurbished & Recertified Pilots W/Platinum Alloyed Valve Discs
| Plant = 
| Reporting criterion = 
| Power level = 
| Mode = 
| Docket = 05000341
| LER year = 1996
| LER number = 17
| LER revision = 0
| Event date = 
| Report date = 
| ENS = 
| abstract = 
}}

=text=
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On October 25,1996, during the fifth refueling outage (RF05), Detroit Edison determined that only one of the HReen safety relief valve (SRV) pilot assemblies tested had lined within the technical specification (TS) one percent allowable tolerance. Six of the fineen SRV pilots tested failed to lin within the pressure limit of the test system.
The sause of the SRV setpoint drin has been continued to be oxide bonding of the pilot valve disc and l
seat. Detroit Edison believes the disc bonding was exacerbated by the following identified Fermi 2 Cycle 5 conditions: (1) Frequent operation at lower power levels may have resulted in 10, er quality steam (higher moisture content) than in previous cycles; (2) There were more cooldown and heatup cycles experienced in Cycle 5 than in Cycle 3 and Cycle 4; and (3) Cycle 5 was 597 at power days, resulting in a longer time for the oxide bond to suengthen. Furthermore, Detroit Edison believes the Cycle 5 results may have been affected by the C ele 5 initial setpoint lift testing method which used nitrogen at ambient 3
room temperature.
Corrective actions include: all fineen Cycle 5 pilots were replaced during the extended finh refueling outage with refurbished and recertified pilots with platinum alloyed valve discs; these fifteen refurbished SRV pilot valve assemblies with platinum alloyed discs were replaced with pilots containing either Stellite or platinum alloyed dises, and tested during a planned Cycle 6 mid-cycle outage; a License Amendment request to modify the SRV setpoint tolerance criteria from i 1% to 1 3% will be submitted to the NRC, and it is intended that platinum ion beam bombarded discs will be installed during the next refueling outage.
9711180231 971110 PDR ADOCK 05000341 8
PDR

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LICENSEE EVENT REPORT (LER) TEXT CONTINUATION
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2 l 0' l 9 Form 12 0 l5 l0 l0l0 l3l4l1 96-0 1 7 m,n Initial Plant Condition:
Operational Condition:
5 (Refueling)
Reactor Power:
0 Percent Reactor Pressure:
O psig j
Reactor Temperature:
98 degrees Fahrenheit Descriotion of the Event:
I A.

===Background===
The) fain Steam System [SB] is equipped with fifteen Target Rock two stage pilot operated Safety Relief Valves (SRVs)[AC][RVJ. The safety function of the SRVs is to prevent the reactor coolant system from being pressurized to more than 110 percent (1375 psig) of the reactor pressme vessel (RPV)[ AC][RPVJ design pressure of 1250 psig in accordance with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section Ill, 1968, Nuclear Vessels. The Ferm12 Technical Specificatlons (TS) Bases state a Safety Limit of 1325 psig vessel steam dome pressure to ensure the RPV bottom pressure does not exceed the 1375 psig value. The Fermi 2 Updated Final Safety Analysis Report (UFSAR) and corresponding General Electric (GE) reload licensing overpressure analysis demonstrate that a total of eleven operable SRVs are adequate to limit reactor pressure to less than ASME B&PV Code allowable values for the worst case transient.
The Target Rock SRV is a dual function valve which can be actuated by system steam pressure exceeding the pilot disc setpoint spring force (the Code safety mode), or it can be actuated by an electrical signal to its electro-pneumatic actuator (relief mode). The electro pneumatic actuator removes the setpoint spring force so that full steam line pressure lifts the pilot disc. The relief mode of operation was installed for manual operation and for RPV automatic depressurization system (ADS) operation (the nuclear safety-related ftmetion applicable to five of the fifteen SRVs). Two additional SRVs receive signals from an automatic control logic using relief mode actuation for the low-low-set (LLS) function that was added prior to initial licensing of Fermi 2.
TS 3.4.2.1 requires that for the safety valve function, at least eleven of the following SRVs shall be operable with the specified code safety valve function lift settings during Operational Conditions 1,2 and 3:
5 safety / relief valves at 113f psig +/ 1 percent 5 sakty/ relief valves at 1145 psig +/- 1 percent 5 safety / relief valves at 1155 psig +/- 1 percent TS Surveillance 4.4.2,1 requires that at least half of the SRVs be set pressure tested at least once every eighteen months, such that all fifteen SRVs are set pressure tested at least once every forty months. Detroit Edison, based on past results, tests all fifteen SRVs each refueling outage.
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==Event Description==
During the fifth refueling outage, the SRV pilot valves were being tested in accordance with TS 4.4.2.1 using a nitrogen testing system at ambient room temperature. This initial testing identified one pilot valve that passed within the one percent acceptance criteria and two pilot
. valves that would not lifl under the maximum applied testing system pressure. Based on an inspection of the pilot discs of these two pilot valves, oxide bonding was determined to be the l
l cause for the failure of the pilot valve assemblics to actuate. In addition, one SRV pilot valve assemb_ly was damaged during installation into the testing system and could not be immediately tested. The remainder of the pilot valves were shipped oft site for confirmatory, independent, testing.
On October 25,1996, Detroit Edison determined that eleven of the filleen pilot valves would not have lified within the TS one percent allowable tolerance. Subsequent testing determined that of the filleen pilot valves, six would not open under maximum applied test pressure. On November 9,1996, aller consultation with Target Rock, a special test was performed at Wyle Laboratories on the damaged pilot valve that demonstrated the ability of the pilot to lifl. Following is a table
.ummarizing the results of the testing:
Valve '
Nominal As Found Percent Number Setnoint (psig)
Setnoint (psig)
Drib 32104 F013A 1135 +/- 11.3
>l2R8 B2104 F01311 1135 +/- 11,3 1230
+ 8.3 132104 F013C 1135 +/- 11.3 1172
+ 3.3 112104 F0130-1135 +/ 11.3 1186
+ 4.5 B2104 F013K 1135 +/- 11.3
>1273 B2104 F013D 1145 +/- !!.4
>l300 B2104 F013F 1145 +/- 11.4 1218
+ 6.4 B2104 F013L 1145 +/- 11,4 1245
+ 8.7 B2104 F013M
:- 1145 +/ 11.4 1170
+ 2.2 B2104 F01.,
i145 +/- 11.4 1221
+ 6.6 B2104 F013E 1155 +/- 11.5
>1282 B2104 F01311 1155 +/ 11.5 1145
:- 0.8 B2104 F013J l155 +/- 11.5
>l306 B2104 F013P 1155 +/- 11.5
>l376 B2104 F013R 1155 +/ 11.5 1179 **
+ 2.1
* Did not open during testing.
** Damaged pilot assembly successfully tested on November 9,1996.

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LICENSEE EVENT REPORT (LER) TEXT CONTINUATION
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lWl9 Fermi 2 0 l5 l0 l0l0 l3l4 l1 96 0 1 7 TL AT Ps All fineen pilot assemblics were replaced during RFOS with refurbished and recertified assemblies, using platinum alloyed discs, for the first portion of Cycle 6 plant operation. A mid-cycle 6 outage was conducted in October 1997, during which the platinum alloyed disc pilots were removed and tested at Wyle Laboratories. All fineen pilots lifted on the first attempt; however, only three of the j
fineen pilot assemblics were found to be within the required il% setpoint tolerance. Following is a table summarizing the results of the testing:
Valve Nominal As Found Percent Number Setnoint (psig)
Setnoint (psig)
LhiD t
il2104 F013A -
1135 +/ 11.3 1148.6 1.2 Il2104 F01311 1135 +/ 11.3 1130.1 0.4 B2104 F013C 1135 +/ 11.3 1152.7 1.6 il2104 F0130 1135 +/- 11.3 1191.0 4.9 il2104 F013K 1135 +/ 11.3 1153.8 1.7 112104 F013D 1145 +/ 11.4 1195.7 4.4 112104 F013F 1145 +/ 11.4 1224.1 6.9 l
112104 F0131, 1145 +/ 11.4 1170.9 2.3 112104 F013M i 145 +/- 11.4 1160.5 1.4 Il2104 F013N 1145 +/- 11.4 1165.8 1.8 112104 F013E 1155 +/- 11.5 1124.6 2.6 112104 F01311 1155 +/ 11.5 1193.6 3.3 Il2104 F013J l 155 +/ 11.5 1177.9 2.0 112104 F013P 1155 +/ 11.5 1148.6 0.6 112104 F013R I155 +/- 11.5 1156.6 0.1

==Cause of the Event==
Industry studies perfonned to date on the Target Rock two stage SRV have indicated that the cause for the high set point drift test results are attributable to either binding in the labyrinth seal area caused by
. tolerance buildup during manufacturing, or dise to seat corrosion bonding caused by oxides of the disc and seat material ibnning singlejoined matrices and inhibiting disc movement. Based on previous Fermi 2 inspection and the testing perfonned at Wyle Laboratories (setpoint returning to nonnal aner initial operation of the pilot valve, thereby breaking the bond), labyrinth seal binding has been climinated as a failure mechanism for the Cycle 5 high setpoint drin test results, llased on independent analysis, Detroit Edison has concinded that the corrosion bonding is the predominant cause for the high setpoint test results. For the corrosion bonding condition, oxygen, temperature and moisture are required to funn the bond. The initial oxide bond is believed to form relatively quickly, and then strengthens over time as a result of stellite cobalt diflusion and plant specific steam moisture and chemical content. Detroit Edison believes that a significant contributor to
:- the magnitude of the oxide bond experienced for Cycle 5 was the 597 at power day operating cyce, n.
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TUT dire resulting in a longer time for the oxide bond to strengthen (Cycle 4 was 401 at power days, Cycle 3 was 456 at power days).
Detroit Edison has investigated the plant operational differences between Cycle 5 and previous cycles and believes the corrosion bonding was exacerbated by the following Cycle 5 identified conditions:
(1) Frequent operation at low power levels due to turbine vibration and hydrogen cooler problems, as well as several forced outages may have resulted in lower quality steam (higher moisture content)in the SRV piping area; and (2) the number of cooldown and heatup cycles for Cycle 5 was greater than for Cycle 3 and Cycle 4.
Chemistry for Cycle 5 was also eve.luated. No unusual excursions from previous cycles were identified that could account for the degree of SRV pilot setpoint drift due to the oxide bo id formation. Offsite laboratory analysis of the composition of the oxide on the disc to seat bonds did not identify any elements present that are not consistent with 'ndustry analyzed oxide bonding. Therefore, Detroit Edison does not believe that Fermi 2 Cycle 5 chemistry was a significant contributor to the oxide bond formation.
A review of SRV location and pilot valve serial number was performed. No correlation between pilot valve. serial number, tail pipe temperatures recorded (indication of SRV leakage), steam line location, setpoint grouping nor function (ADS or LLS) was identified.
The records for the previous two pilot valve refurbishment and recertification processes were reviewed. The pilot valves were refurbished in accordance with Target Rock procedures and recertified in accordance with approved test facility procedures under the supervision of Detroit Edison personnel with no significant anomalies identified.
The results from the mid Cycle 6 outage realiinn the conclusions stated above regarding the early fomtation of the oxide bond and its strenghtening over time. The first several months of Cycle 6 consisted of a series of startups and shutdowns, with t much continuous run time. The reactor was then operated for approximately six months at near fun power before it was shut down for the mid cycle outage in October 1997. The platinum alloyed pilots were installed for over 300 days during this part of Cycle 6, approximately 180 of which were at power, allowing the oxide bond to fomi but not strenghten significantly, in early 1997. during the period the platinum alloyed discs were installed, the BWROO concluded that the platinum alloyed discs were not significantly better than the Stellite discs at resisting the oxide bond formation and that the industry should not pursue this fix any longer. This conclus:on was based on industry experience and was presented to the industry and to the NRC in a teleconference on May 7,1997. Ti crefore, the filleen platinum alloyed dise pilots were replaced in the mid-Cycle 6 outage with the twelve Stellite and three platinum alloyed disc pilots that were available in the Fenni 2 warehouse.

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6 l0'l9 Fermi 2 0 lS l0 l0l0 l3l4l1 96 0 1 7 Ti si p h

==Analysis of the Event==
The SRV capacity of the Fermi 2 plant is sized to limit the primary system pressure, including transients, to the requirements of the AShiE B&PV Code, Section 111,1968, Nuclear Vessels.
Pressure relieving devices may not alone provide complete protection without crediting other protective functions as permitted by the AShiE Code. The original Fermi 2 SRV sizing evaluation, described in the Fermi 2 UFSAR, assumes credit for operation of the scram protective system. Fermi 2 reload analysis of the reactor vessel overpressure protection capacity is currently perfomied by GE, Nuclear Energy Division, using their licensed GESTAR 11 methodology, Under the general requirements for protection against overpressure as given in Section ill of the AShiE Code, credit can be allowed for a scram from the reactor protection system. The backup reactor high neutron flux scram is conservatively applied as a basis for determining the required capacity of the pressure relieving dual purpose SRVs. The application of the direct position scrams could be used since they qualify as acceptable pressure protection devices when determining the required SRV capacity under the provisions of the AShiE Code. The direct position scram circuitry is required to be operable in accordance with TS 3.1.1. The operability of the hiSIV position switches is routinely proven by TS channel functional testing on a quarterly basis during the operating cycle. The full logic functional test of each channel, from each hiSIV switch to and including the relay which initiates the protection system trip, is perfonned during each refueling outage.
The SRV capacity required for overpressure protection is determined based upon the minimum capacity that will provide an adequate margin between the vessel design pressure and the 110 percent vessel AShiE code limit in response to the hiain Steam isolation Valve (htSIV) closure Ilux scram event. The required SRV capacity is determined by analyzing the pressure rise from an hiSIV closure with a flux scram transient. The reactor was assumed to be operating at a maximurn TS vessel dome pressure of 1045 psig. The analysis hypothetically assumed the failure of the direct isolation valve position scram. For the analysis, relief setpoints of the eleven credited SRVs are assumed to be in the range of approximately 1169 to 1190 psig (103 percent of the nominal setpoint).
Ilowever, accordmg to standard GE practice (for original sizing and reload licensing analysis), the SRV capacity evaluation does not assume credit for the direct scram, only the indirect flux scram is assumed. Further, no credit is taken for the externally powered relief mode of the dual purpose SRVs in their ash 1E Code qualified mode of safety operation. The basis of this GE practice is that the reactor protection system is assumed to be subject to a component, logic, or system failure such that one of the trip signals fails, Thus, althour,h the potential failure of the htSIV position switches is of low probability, it is conservatively assumed to fail for the licensing analyses perfonned prior to initial or reload cycles.
The reactor protection system scram may be tripped by any one of three sources; i.e., direct, neutron flux, or pressure signal. The direct scram signal is derived from position switches mounted on the main steam line isolation valves or the turbine stop valves or from picoute switches mounted on the dump valve of the turbine control valve hydraulic actuation synem. The hiSIV closure position

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION
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l"l9 Fermi 2 0 l5 l0 l0l0 l3l4 l1 96 0 1 7 switches are actuated when a closure of the control valves is initiated and following eight percent of full stroke hiSIV closure travel (TS setroint).
The logic functional tests successfully performed during the fillh operating cycle and duri.)g the fillh refueling outage demonstrate that the hiSiv position scram has been effective throughout Cycle 5.
This meets the condition for applying the reactor protective system scram credit as allowed by the AShiE Code to an overpressure analysis of h1SIV closure with direct scram for Cycle 5.
The Cycle 5 reload licensing analyses for Fermi 2 were perfonned assuming inoperability of the four lowest setpoint SRVs, with the remaining eleven SRVs having set pressures three percent above nominal. The most severe overpressure event resulting from these assumptions yields a peak dome pressure of 1307 psig and a peak RPV bottom pressure of 1323 psig, both of which are within acceptance ciiteria. Using the as fbund condition for all fifteen of the SRVs (six failed, one pass, eight at various setpoints greater than one percent), the most severe overpressure event resulting from these assumptions yields a peak dome pressure of 1349 psig and a peak bottom pressure of 1366 psig.
The overpressure analysis conducted as noted above produces results within the Code acceptance criterion. Ilowever, Fenni 2 has also evaluated two additional overpressure events based upon the infonnation now available from Cycle 5. Using a turbine trip without bypass with a reactor scram on turbine stop valve position (more severe than the all hiSIV closure event with scram on h1SIV position) from 102 percent of rated power, and all other reload licensing assumptions, this event would yield a peak dome pressure of 1262 psig and a peak RPV bottom pressure of 1283 psig, which are within the Fenni 2 TS Safety Limit and the Ash 1E B&PV Code, Section 111, acceptance criteria of 1325 psig and 1375 psig, respectively. The all h1SIV closure event using the direct h1SIV position as the reactor scram initiation signal yields a peak dome pressure of 1241 psig and a peak bottom pressure of 1263 psig. This analysis satisfies both the AShiE Code and the TS Safety Limit.
Therefbre, the health and safety of the public were not adversely impacted by the condition of the SRV pilots at the end of Fenni 2 Cycle 5.
The data from the SRV pilots removed during the mid Cycle 6 outage show significant improvement over that obtained from the Cycle 5 SRVs. The mid-Cycle 6 results cre easily bounded by the SRV peribrmance assumptions used in the Cycle 6 Reload Licensing Submittal, and thus, the health and safety of the public also were not adversely affected by the SRV conditions existing during the first portion of Cycle 6.
Detroit Edison believes that operation for the remainder of Cycle 6 does not have the potential to adversely impact the health and safety of the public. The SRV pilots in service for the remainder of Cycle 6 will not be subjected to the service conditions believed to exacerbate oxide bonding. The recent turbine upgrades have resulted in better plant perfbnnance with fewer power reductions. In addition, the Cycle 6 reload licensing overpressure analysis has more margin than the Cycle 5 overpressure analysis due to Cycle 6 core perfbnnance characteristics.

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==Corrective Actions==
A.

==Immediate Corrective Actions==
EfD1 1
Twelve pilot valves were sent to an off site vendor for confirmatory testing, including the damaged pilot valve assembly, ne damaged pilot valve stabilizer disc was broken off, and discussion with Target Rock and Wyle Laboratories indicated that this valve could be initial lin tested, but that follow up lin testing could not be conducted (the pilot vah would chatter). The eleven undamaged pilot valves all " popped" within three percent of their nominal pressure setpoint aner their initial lin (achieved either by the pressure setpoint tests or by subsequent relief mode tests).
Six pilot vahes falling to open on demand during testing did not meet the Fermi 2 maintenance rule performance criteria for the reactor pressure vessel (RpV) overpressure protection system.
Therefore, the RpV overpressure protection system has been classified as a 10CFR50.65 Maintenance Rule (a)(1) system requiring that a reliability improvement plan be developed.
All filleen pilot assemblics were replaced with refurbished and recertified assemblies, using platinum alloyed discs, in the SRVs for Cycle 6 plant operation.
Mid Cvele 6 Outane The fineen SRV pilot valve assemblies with platinum alloyed discs installed during the extended finh refueling outage were sent to Wyle Laboratories for setpoint testing. All fineen assemblies
:- were replaced with refurbished assemblies, twelve contr.ining Stellite 6B discs and three containing platinum alloyed discs.
B.
Corrective Actions to Prevent Recunrence
: 1. The fineen refurbished SRV pilot valve assemblies with platinum alloyed discs installed during the extended finh refueling outage will be replaced, using pilots with either Stellite or platinum alloyed discs, and tested during a planned Cycle 6 mid-cycle outage. This action was completed during the October 1997 mid-cycle 6 outage.
: 2. A License Amendment request to modify the SRV setpoint tolerance criteria from i 1% to 1 3% will be submitted to the NRC by January 30,1998.
l
: 3. Detroit Edison will submit writtenjustification to the NRC by January 14,1998 for operating with the currently installed SRV pilot assemblies for the duration of this operating cycle.
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7 TEAT (17; 4.- Detroit Edison intends to install platinum ion beam bombarded discs during the next refueling outage (RFO 6).
i S. Detroit Edison will continue to follow industry developments related to the improved reliability of SRV se' point performance.
Additionalinformation:
A.
Failed Components Component:
hiain Steam Safety Relief Valve

==Description==
Two Stage Safety Relief Valve hianufacturer/hiodel:
Target Rock Company, hiodel 7567F 11.
Previous LERs on Similar Problems LER 94-002:
" Safety Relief Valve Set Pressures Outside of Technical Specification Limits" LER 92 009:
" Safety Relief Valves Set Pressure outside Technical Specification Limit" LER 91-013:
" Safety Relief Valves Set Pressure Outside Technical Specification Limit" LER 89-028:
" Safety Relief Valves Fail Their Set Pressure Tolerance Test" 1
LER 88 009
" Safety Relief Valves Fail Their Set Pressure Tolerance Test" LhR 86 013:
" Reactor Coolant System Safety Relief Valves Exceed Nameplate Set Pressure Surveillance Test Tolerances"
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